Similar Documents at Perry |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARPY-CEI-NRR-2438, Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl1999-10-14014 October 1999 Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl ML20217F3901999-10-14014 October 1999 Discusses Request That Proprietary Document NEDE-32907P,DRF A22-0084-53, Safety Analysis Rept for Perry NPP 5% Power Uprate, Class III Dtd Sept 1999 Be Withheld.Determined Document Proprietary & Will Be Withheld ML20217G9201999-10-14014 October 1999 Discusses Utils Request for Approval of Quality Assurance Program Changes PY-CEI-NRR-2435, Responds to NRC Re Violations Noted in Insp Rept 50-440/99-13.Corrective Actions:Ts SRs with Incorrect Descriptions Were Annotated to Ensure That CR Operators Are Aware That ACs Are Effect1999-10-13013 October 1999 Responds to NRC Re Violations Noted in Insp Rept 50-440/99-13.Corrective Actions:Ts SRs with Incorrect Descriptions Were Annotated to Ensure That CR Operators Are Aware That ACs Are Effect ML20212K9271999-09-30030 September 1999 Refers to 990927 Meeting Conducted at Perry Nuclear Power Plant to Discuss Initiatives in Risk Area & to Establish Dialog Between SRAs & PSA Staff ML20212J1451999-09-30030 September 1999 Forwards Order,In Response to 990505 Application PY-CEI/NRR- 2394L.Order Approves Conforming License Amend Which Will Be Issued & Made Effective When Transfer Completed ML20217E7111999-09-30030 September 1999 Documents Telcon Conducted on 990929 Between M Underwood of Oh EPA & D Tizzan of Pnpp,Re Request to Operate Pnpp Sws,As Is,Until Resolution Can Be Obtained ML20212G4161999-09-24024 September 1999 Informs of Completion of Licensing Action for Generic Ltr 98-01, Y2K Readiness of Computer Systems at Nuclear Power Plants, for Perry Nuclear Power Plant ML20212G5811999-09-23023 September 1999 Informs That Licenses for Ta Lentz,License SOP-31449,PJ Arthur,License SOP-30921-1 & Dp Mott,License SOP-31500 Are Considered to Have Expired,Iaw 10CFR50.74(a),10CFR55.5 & 10CFR55.55 PY-CEI-NRR-2432, Forwards NRC Form 536, Operator Licensing Exam Data, in Response to Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams1999-09-21021 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, in Response to Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams PY-CEI-NRR-2428, Submits Resolution to Seventh Question Proposed within NRC 980615 RAI Relating to Cooling Water Sys That Serve Containment Air Coolers & Assessment,Post Accident,Of Potential Water Hammer & two-phase Flow Conditions1999-09-16016 September 1999 Submits Resolution to Seventh Question Proposed within NRC 980615 RAI Relating to Cooling Water Sys That Serve Containment Air Coolers & Assessment,Post Accident,Of Potential Water Hammer & two-phase Flow Conditions ML20212D0151999-09-14014 September 1999 Requests Cancellation of NPDES Permit 3II00036.Permit Has Been Incorporated in Permit 3IB00016*ED.Discharge Point Sources & Associated Fees Currently Covered Under Permit 3IB00016*ED ML20212A8371999-09-13013 September 1999 Forwards Insp Rept 50-440/99-13 on 990712-30 & Notice of Violation.Insp Included Evaluation of Engineering Support, Design Change & Modification Activities,Internal Assessment Activities & Corrective Actions ML20217A8971999-09-0909 September 1999 Forwards Insp Rept 50-440/99-09 on 990709-0825.One Violation of NRC Requirements Occurred & Being Treated as NCV, Consistent with App C of Enforcement Policy PY-CEI-NRR-2431, Forwards Revised Emergency Plan for Perry NPP, IAW 10CFR50.54(q).Changes Constitute Revs,Temporary Changes or Reissued Pages1999-09-0909 September 1999 Forwards Revised Emergency Plan for Perry NPP, IAW 10CFR50.54(q).Changes Constitute Revs,Temporary Changes or Reissued Pages ML20211Q6911999-09-0606 September 1999 Informs That NRC Tentatively Scheduled Initial Licensing Exam for Perry Operator License Applicants During Wks of 010108 & 15.Validation of Exam Will Occur at Station During Wk of 001218 IR 05000440/19990011999-08-31031 August 1999 Requests That Page Number 4, P2 Status of EP Facilities, Equipment & Resources, of Insp Rept 50-440/99-01 Be Replaced with Encl Rev PY-CEI-NRR-2425, Forwards Copy of Oh EPA Approval for Use of Nalco 7348 & Nalco 7468 at Pnpp,Iaw License NPF-58,App B,Epp,Section 3.21999-08-26026 August 1999 Forwards Copy of Oh EPA Approval for Use of Nalco 7348 & Nalco 7468 at Pnpp,Iaw License NPF-58,App B,Epp,Section 3.2 PY-CEI-NRR-2411, Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl1999-08-19019 August 1999 Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl ML20210Q8421999-08-13013 August 1999 First Final Response to FOIA Request for Documents.Records Listed in App a Being Released in Entirety & Records Listed in App B Being Withheld in Part (Ref FOIA Exempt 5) ML20210R7861999-08-12012 August 1999 Forwards Insp Rept 50-440/99-12 on 990712-16.No Violations Noted.New Emergency Preparedness Program Staff & Mgt Personnel Were Professional & Proactive ML20210S3961999-08-11011 August 1999 Requests That Ten Listed Individuals Be Registered to Take 991006 BWR Gfes of Written Operating Licensing Exam.Two Listed Personnel Will Have Access to Exams Before Exams Are Administered PY-CEI-NRR-2423, Provides Final Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Npps. Remediation of Meteorological Monitoring Sys Has Been Completed & Ppnp Facility Is Now Y2K Ready1999-08-10010 August 1999 Provides Final Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Npps. Remediation of Meteorological Monitoring Sys Has Been Completed & Ppnp Facility Is Now Y2K Ready PY-CEI-NRR-2421, Forwards Semiannual fitness-for-duty Rept,Iaw 10CFR26.71(d) for Pnpp Covering Period of 990101-9906301999-08-10010 August 1999 Forwards Semiannual fitness-for-duty Rept,Iaw 10CFR26.71(d) for Pnpp Covering Period of 990101-990630 ML20210Q7981999-08-10010 August 1999 Informs That Intention to Utilize Sulfuric Acid in Pnpp Circulating Water Sys to Lower Ph Is Anticipated to Be Completed in Nov ML20210Q5831999-08-10010 August 1999 Requests Permission to Chemically Treat CWS for Algae Due to Cooling Water Basin Becoming Infected with Algae of Various Types PY-CEI-NRR-2422, Forwards Addl Info Re ASME Section IX Relief Request (IR-023) for Inservice Examination Program at Pnpp,Submitted on 9808261999-08-10010 August 1999 Forwards Addl Info Re ASME Section IX Relief Request (IR-023) for Inservice Examination Program at Pnpp,Submitted on 980826 ML20210K6331999-08-0404 August 1999 Submits Response to Requests for Addl Info to GL 92-01,Rev 1, Reactor Vessel Structural Integrity, for Perry Nuclear Plant,Unit 1 PY-CEI-NRR-2417, Forwards NP0059-007, Pnpp - Unit 1 ISI Summary Rept Results for Outage 7 (1999) First Period,Second Interval, IAW ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,Article IWA-60001999-08-0202 August 1999 Forwards NP0059-007, Pnpp - Unit 1 ISI Summary Rept Results for Outage 7 (1999) First Period,Second Interval, IAW ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,Article IWA-6000 ML20210G8411999-07-28028 July 1999 Informs That Based on Determination That Rev 14 Changes to Various Portions of Perry Nuclear Power Plant Emergency Plan Does Not Decrease Effectiveness of Licensee Emergency Plan & Meets Standard of 10CFR50.47(b),no NRC Approval Required ML20210E0661999-07-22022 July 1999 Forwards Insp Rept 50-440/99-08 on 990518-0708.No Violations Noted.Overall Conduct of Activities at Perry Facility, Conservative & Professional with Continuing Focus on Safety PY-CEI-NRR-2419, Forwards Four Copies of Rev 27 to Pnpp Security Plan,Per 10CFR50.54(p)(2).Changes Have Been Determined Not to Decrease Effectiveness of Security Plan.Rev Withheld1999-07-21021 July 1999 Forwards Four Copies of Rev 27 to Pnpp Security Plan,Per 10CFR50.54(p)(2).Changes Have Been Determined Not to Decrease Effectiveness of Security Plan.Rev Withheld PY-CEI-NRR-2415, Submits Estimate of Number of Licensing Submittal for FY00 & 01,for Pnpp,Per Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates1999-07-19019 July 1999 Submits Estimate of Number of Licensing Submittal for FY00 & 01,for Pnpp,Per Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates ML20210A6441999-07-15015 July 1999 Advises That Listed Operator Licenses for Company Personnel Have Expired,As of 990715,per 10CFR50.74(a) & 10CFR55.5. Individuals Listed Have Assumed Responsibilities at Pnpp That Do Not Require Operator Licenses ML20209E5951999-07-0909 July 1999 Ltr Contract:Task Order 46, Perry Engineering & Technical Support (E&Ts) Insp, Under Contract NRC-03-98-021 ML20209D5931999-07-0101 July 1999 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Investigation Rept 3-98-005 Issued on 990510.Corrective Actions Will Be Examined During Future Inspections ML20211A3881999-06-30030 June 1999 Supplements Re NRC Incomprehensible Enforcement Inactions.Understands That NRC Claiming Hands Tied in Matter Re Discrimination Violation at Perry Npp.Requests That NRC Take Listed Two Actions PY-CEI-NRR-2410, Submits Response to NRC Request for Info Re GL 98-01, Y2K Readiness of Computer Sys at Npps. Attachment 1 Provides Y2K Readiness Disclosure for Plant1999-06-29029 June 1999 Submits Response to NRC Request for Info Re GL 98-01, Y2K Readiness of Computer Sys at Npps. Attachment 1 Provides Y2K Readiness Disclosure for Plant PY-CEI-NRR-2413, Advises That Jk Wood Will Replace Lw Myers as Pnpp,Vice President - Nuclear,Effective 9907061999-06-29029 June 1999 Advises That Jk Wood Will Replace Lw Myers as Pnpp,Vice President - Nuclear,Effective 990706 PY-CEI-NRR-2403, Submits Response to RAI Re GL 96-05 Program at Plant1999-06-29029 June 1999 Submits Response to RAI Re GL 96-05 Program at Plant ML20196H9641999-06-29029 June 1999 Confirms 990615 Telcon Request with J Lieberman for Addl Time to Respond to Enforcement Action 99-012.FirstEnergy Has 60 Days to Respond to EA PY-CEI-NRR-2412, Forwards Perry Nuclear Plant Simulator Four Yr Test Rept & NRC Form 474, Simulation Facility Certification, Which Describes Tests Conducted from June 1995-19991999-06-28028 June 1999 Forwards Perry Nuclear Plant Simulator Four Yr Test Rept & NRC Form 474, Simulation Facility Certification, Which Describes Tests Conducted from June 1995-1999 ML20211A3991999-06-28028 June 1999 Submits Concerns Re Proposed Changes to Enforcement Policy. Believes That NRC Existing Regulations & Policies Provide Adequate Controls & That NRC Not Administering Consistent Enforcement Policies in Response to Discriminatory Actions ML20196A6601999-06-16016 June 1999 Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1 PY-CEI-NRR-2405, Informs NRC That Pnpp Staff Has Implemented Formal Industry Position on Severe Accident Mgt1999-06-14014 June 1999 Informs NRC That Pnpp Staff Has Implemented Formal Industry Position on Severe Accident Mgt ML20195H9051999-06-10010 June 1999 Forwards Two License Renewal Applications Consisting of NRC Form 398 & NRC Form 396 for Bk Carrier,License OP-30997 & Jt Steward,License OP-30564-1 ML20212H5841999-06-10010 June 1999 Responds to NRC Notice of Level III Violation Received by K Wierman.Corrective Actions:K Wierman Has Been Dismissed from Position at Perry NPP PY-CEI-NRR-2408, Forwards Rev 4 to Pump & Valve Inservice Testing Program Plan, Incorporating Changes to Program to Implement Second Ten Yr Insp Interval1999-06-10010 June 1999 Forwards Rev 4 to Pump & Valve Inservice Testing Program Plan, Incorporating Changes to Program to Implement Second Ten Yr Insp Interval ML20195G6741999-06-10010 June 1999 Forwards Insp Rept 50-440/99-03 on 990407-0517.Two Violations of NRC Requirements Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy PY-CEI-NRR-2407, Responds to NRC Re Violations Noted in Investigations Rept 3-98-005.Corrective Actions:Initiated Internal Investigation Into Matter to Discover Party Responsible for Falsification of Training Records1999-06-0909 June 1999 Responds to NRC Re Violations Noted in Investigations Rept 3-98-005.Corrective Actions:Initiated Internal Investigation Into Matter to Discover Party Responsible for Falsification of Training Records 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARPY-CEI-NRR-2438, Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl1999-10-14014 October 1999 Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl PY-CEI-NRR-2435, Responds to NRC Re Violations Noted in Insp Rept 50-440/99-13.Corrective Actions:Ts SRs with Incorrect Descriptions Were Annotated to Ensure That CR Operators Are Aware That ACs Are Effect1999-10-13013 October 1999 Responds to NRC Re Violations Noted in Insp Rept 50-440/99-13.Corrective Actions:Ts SRs with Incorrect Descriptions Were Annotated to Ensure That CR Operators Are Aware That ACs Are Effect ML20212G5811999-09-23023 September 1999 Informs That Licenses for Ta Lentz,License SOP-31449,PJ Arthur,License SOP-30921-1 & Dp Mott,License SOP-31500 Are Considered to Have Expired,Iaw 10CFR50.74(a),10CFR55.5 & 10CFR55.55 PY-CEI-NRR-2432, Forwards NRC Form 536, Operator Licensing Exam Data, in Response to Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams1999-09-21021 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, in Response to Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams PY-CEI-NRR-2428, Submits Resolution to Seventh Question Proposed within NRC 980615 RAI Relating to Cooling Water Sys That Serve Containment Air Coolers & Assessment,Post Accident,Of Potential Water Hammer & two-phase Flow Conditions1999-09-16016 September 1999 Submits Resolution to Seventh Question Proposed within NRC 980615 RAI Relating to Cooling Water Sys That Serve Containment Air Coolers & Assessment,Post Accident,Of Potential Water Hammer & two-phase Flow Conditions ML20212D0151999-09-14014 September 1999 Requests Cancellation of NPDES Permit 3II00036.Permit Has Been Incorporated in Permit 3IB00016*ED.Discharge Point Sources & Associated Fees Currently Covered Under Permit 3IB00016*ED PY-CEI-NRR-2431, Forwards Revised Emergency Plan for Perry NPP, IAW 10CFR50.54(q).Changes Constitute Revs,Temporary Changes or Reissued Pages1999-09-0909 September 1999 Forwards Revised Emergency Plan for Perry NPP, IAW 10CFR50.54(q).Changes Constitute Revs,Temporary Changes or Reissued Pages PY-CEI-NRR-2425, Forwards Copy of Oh EPA Approval for Use of Nalco 7348 & Nalco 7468 at Pnpp,Iaw License NPF-58,App B,Epp,Section 3.21999-08-26026 August 1999 Forwards Copy of Oh EPA Approval for Use of Nalco 7348 & Nalco 7468 at Pnpp,Iaw License NPF-58,App B,Epp,Section 3.2 PY-CEI-NRR-2411, Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl1999-08-19019 August 1999 Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl ML20210S3961999-08-11011 August 1999 Requests That Ten Listed Individuals Be Registered to Take 991006 BWR Gfes of Written Operating Licensing Exam.Two Listed Personnel Will Have Access to Exams Before Exams Are Administered PY-CEI-NRR-2421, Forwards Semiannual fitness-for-duty Rept,Iaw 10CFR26.71(d) for Pnpp Covering Period of 990101-9906301999-08-10010 August 1999 Forwards Semiannual fitness-for-duty Rept,Iaw 10CFR26.71(d) for Pnpp Covering Period of 990101-990630 PY-CEI-NRR-2423, Provides Final Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Npps. Remediation of Meteorological Monitoring Sys Has Been Completed & Ppnp Facility Is Now Y2K Ready1999-08-10010 August 1999 Provides Final Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Npps. Remediation of Meteorological Monitoring Sys Has Been Completed & Ppnp Facility Is Now Y2K Ready PY-CEI-NRR-2422, Forwards Addl Info Re ASME Section IX Relief Request (IR-023) for Inservice Examination Program at Pnpp,Submitted on 9808261999-08-10010 August 1999 Forwards Addl Info Re ASME Section IX Relief Request (IR-023) for Inservice Examination Program at Pnpp,Submitted on 980826 PY-CEI-NRR-2417, Forwards NP0059-007, Pnpp - Unit 1 ISI Summary Rept Results for Outage 7 (1999) First Period,Second Interval, IAW ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,Article IWA-60001999-08-0202 August 1999 Forwards NP0059-007, Pnpp - Unit 1 ISI Summary Rept Results for Outage 7 (1999) First Period,Second Interval, IAW ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,Article IWA-6000 PY-CEI-NRR-2419, Forwards Four Copies of Rev 27 to Pnpp Security Plan,Per 10CFR50.54(p)(2).Changes Have Been Determined Not to Decrease Effectiveness of Security Plan.Rev Withheld1999-07-21021 July 1999 Forwards Four Copies of Rev 27 to Pnpp Security Plan,Per 10CFR50.54(p)(2).Changes Have Been Determined Not to Decrease Effectiveness of Security Plan.Rev Withheld PY-CEI-NRR-2415, Submits Estimate of Number of Licensing Submittal for FY00 & 01,for Pnpp,Per Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates1999-07-19019 July 1999 Submits Estimate of Number of Licensing Submittal for FY00 & 01,for Pnpp,Per Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates ML20210A6441999-07-15015 July 1999 Advises That Listed Operator Licenses for Company Personnel Have Expired,As of 990715,per 10CFR50.74(a) & 10CFR55.5. Individuals Listed Have Assumed Responsibilities at Pnpp That Do Not Require Operator Licenses ML20211A3881999-06-30030 June 1999 Supplements Re NRC Incomprehensible Enforcement Inactions.Understands That NRC Claiming Hands Tied in Matter Re Discrimination Violation at Perry Npp.Requests That NRC Take Listed Two Actions PY-CEI-NRR-2403, Submits Response to RAI Re GL 96-05 Program at Plant1999-06-29029 June 1999 Submits Response to RAI Re GL 96-05 Program at Plant PY-CEI-NRR-2413, Advises That Jk Wood Will Replace Lw Myers as Pnpp,Vice President - Nuclear,Effective 9907061999-06-29029 June 1999 Advises That Jk Wood Will Replace Lw Myers as Pnpp,Vice President - Nuclear,Effective 990706 PY-CEI-NRR-2410, Submits Response to NRC Request for Info Re GL 98-01, Y2K Readiness of Computer Sys at Npps. Attachment 1 Provides Y2K Readiness Disclosure for Plant1999-06-29029 June 1999 Submits Response to NRC Request for Info Re GL 98-01, Y2K Readiness of Computer Sys at Npps. Attachment 1 Provides Y2K Readiness Disclosure for Plant ML20211A3991999-06-28028 June 1999 Submits Concerns Re Proposed Changes to Enforcement Policy. Believes That NRC Existing Regulations & Policies Provide Adequate Controls & That NRC Not Administering Consistent Enforcement Policies in Response to Discriminatory Actions PY-CEI-NRR-2412, Forwards Perry Nuclear Plant Simulator Four Yr Test Rept & NRC Form 474, Simulation Facility Certification, Which Describes Tests Conducted from June 1995-19991999-06-28028 June 1999 Forwards Perry Nuclear Plant Simulator Four Yr Test Rept & NRC Form 474, Simulation Facility Certification, Which Describes Tests Conducted from June 1995-1999 ML20196A6601999-06-16016 June 1999 Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1 PY-CEI-NRR-2405, Informs NRC That Pnpp Staff Has Implemented Formal Industry Position on Severe Accident Mgt1999-06-14014 June 1999 Informs NRC That Pnpp Staff Has Implemented Formal Industry Position on Severe Accident Mgt ML20212H5841999-06-10010 June 1999 Responds to NRC Notice of Level III Violation Received by K Wierman.Corrective Actions:K Wierman Has Been Dismissed from Position at Perry NPP ML20195H9051999-06-10010 June 1999 Forwards Two License Renewal Applications Consisting of NRC Form 398 & NRC Form 396 for Bk Carrier,License OP-30997 & Jt Steward,License OP-30564-1 PY-CEI-NRR-2408, Forwards Rev 4 to Pump & Valve Inservice Testing Program Plan, Incorporating Changes to Program to Implement Second Ten Yr Insp Interval1999-06-10010 June 1999 Forwards Rev 4 to Pump & Valve Inservice Testing Program Plan, Incorporating Changes to Program to Implement Second Ten Yr Insp Interval PY-CEI-NRR-2407, Responds to NRC Re Violations Noted in Investigations Rept 3-98-005.Corrective Actions:Initiated Internal Investigation Into Matter to Discover Party Responsible for Falsification of Training Records1999-06-0909 June 1999 Responds to NRC Re Violations Noted in Investigations Rept 3-98-005.Corrective Actions:Initiated Internal Investigation Into Matter to Discover Party Responsible for Falsification of Training Records PY-CEI-NRR-2404, Submits Rept of Reduction in PCT Greater than 50 F Due to Implementation of Ge,Safr/Gestr LOCA Methodology for Fuel Cycle 8.Methodology Replaces Original LOCA Methodology1999-06-0707 June 1999 Submits Rept of Reduction in PCT Greater than 50 F Due to Implementation of Ge,Safr/Gestr LOCA Methodology for Fuel Cycle 8.Methodology Replaces Original LOCA Methodology ML20195E4561999-05-25025 May 1999 Submits Petition Per Other Actions Provision of 10CFR2.206,requesting That Radiation Protection Manager at Perry NPP Be Banned by NRC from Participation in Licensed Activities at & for Any NPP for at Least Five Yrs PY-CEI-NRR-2401, Forwards Changes to ERDS Data Point Library & Communication Survey for Perry Nuclear Power Plant.Changes Are Identified by Margin Rev Bars1999-05-20020 May 1999 Forwards Changes to ERDS Data Point Library & Communication Survey for Perry Nuclear Power Plant.Changes Are Identified by Margin Rev Bars PY-CEI-NRR-2402, Notifies of Completion of Plant Core Shroud Insp in Accordance with GL 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Bwrs,1999-05-18018 May 1999 Notifies of Completion of Plant Core Shroud Insp in Accordance with GL 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Bwrs, PY-CEI-NRR-2394, Forwards Application for Order & Conforming License Amend for Transfer of Interest in OL NPF-58 for Pnpp.Authorization of Ceic to Possess Dl 13.74% Undivided Ownership in Pnpp, Requested1999-05-0505 May 1999 Forwards Application for Order & Conforming License Amend for Transfer of Interest in OL NPF-58 for Pnpp.Authorization of Ceic to Possess Dl 13.74% Undivided Ownership in Pnpp, Requested PY-CEI-NRR-2397, Forwards Revised COLR Cycle 8, & Rev 1 to J11-03371SRLR, Supplemental Reload Licensing Rept for Pnpp,Unit 1 Reload 7,Cycle 8, IAW TS Section 5.6.51999-04-30030 April 1999 Forwards Revised COLR Cycle 8, & Rev 1 to J11-03371SRLR, Supplemental Reload Licensing Rept for Pnpp,Unit 1 Reload 7,Cycle 8, IAW TS Section 5.6.5 PY-CEI-NRR-2396, Provides post-examination Results of Weld Overlay Repair for FW Nozzle to safe-end Weld at Pnpp.Info Supplements Licensee 990325 & 0401 Ltrs Submitted to NRC That Requested Approval of Repair Plan for 1B13-N1C-KB FW Nozzle to safe-en1999-04-28028 April 1999 Provides post-examination Results of Weld Overlay Repair for FW Nozzle to safe-end Weld at Pnpp.Info Supplements Licensee 990325 & 0401 Ltrs Submitted to NRC That Requested Approval of Repair Plan for 1B13-N1C-KB FW Nozzle to safe-end Weld 05000440/LER-1999-002, Forwards LER 99-002-00, LCO 3.0.3 Entered Due to TS Bases Statement Interpretation. Action Occurred After RHR Pump Failed to Start.Plant at No Time Was in Condition Where Required Decay Heat Removal System Were Not Available1999-04-26026 April 1999 Forwards LER 99-002-00, LCO 3.0.3 Entered Due to TS Bases Statement Interpretation. Action Occurred After RHR Pump Failed to Start.Plant at No Time Was in Condition Where Required Decay Heat Removal System Were Not Available PY-CEI-NRR-2388, Forwards Annual Environ & Effluent Release Rept for Pnnp Unit 1 for 1998. Pnpp Had No Liquid Effluent Discharges in 1998 & All Other Release Points Were Well Below Applicable Limits1999-04-22022 April 1999 Forwards Annual Environ & Effluent Release Rept for Pnnp Unit 1 for 1998. Pnpp Had No Liquid Effluent Discharges in 1998 & All Other Release Points Were Well Below Applicable Limits PY-CEI-NRR-2371, Provides Suppl Response to Violations Noted in Insp Rept 50-440/98-18.Corrective Actions:Two Condition Rept Investigations Were Performed to Evaluate Issues Re Fire Protection Program & Addl Training Was Conducted1999-04-21021 April 1999 Provides Suppl Response to Violations Noted in Insp Rept 50-440/98-18.Corrective Actions:Two Condition Rept Investigations Were Performed to Evaluate Issues Re Fire Protection Program & Addl Training Was Conducted PY-CEI-NRR-2382, Forwards 1998 Annual Rept for Firstenergy Corp, for Perry Nuclear Power Plant & Davis-Besse Nuclear Power Station.Form 10-K Annual Rept to Us Securities & Exchange Commission for FY98 Also Encl1999-04-21021 April 1999 Forwards 1998 Annual Rept for Firstenergy Corp, for Perry Nuclear Power Plant & Davis-Besse Nuclear Power Station.Form 10-K Annual Rept to Us Securities & Exchange Commission for FY98 Also Encl PY-CEI-NRR-2392, Forwards Diskette Containing Pnpp Thermoluminescent Dosimeter Doses in Format Requested by Reg Guide 8.7.Without Diskette1999-04-19019 April 1999 Forwards Diskette Containing Pnpp Thermoluminescent Dosimeter Doses in Format Requested by Reg Guide 8.7.Without Diskette PY-CEI-NRR-2387, Documents Telcon Held Between NRC Staff & Members of Perry Nuclear Power Plant Staff on 990406,re Safety Evaluation for License Amend 1051999-04-14014 April 1999 Documents Telcon Held Between NRC Staff & Members of Perry Nuclear Power Plant Staff on 990406,re Safety Evaluation for License Amend 105 PY-CEI-NRR-2386, Forwards Copy of Application Submitted to FERC Proposing Transfer of Jurisdictional Transmission Facilities of Firstenergy Corp Operating Companies to American Transmission Sys,Inc.With One Oversize Drawing1999-04-0606 April 1999 Forwards Copy of Application Submitted to FERC Proposing Transfer of Jurisdictional Transmission Facilities of Firstenergy Corp Operating Companies to American Transmission Sys,Inc.With One Oversize Drawing ML20206N3361999-04-0505 April 1999 Informs NRR of Continued Events Re Transfer of Generation Assets Between Duquesne Light Co & Firstenergy.No Negotiated Settlement Between Firstenergy & Local 270 Pertaining to Generation Asset Swap Has Been Reached PY-CEI-NRR-2384, Provides Supplemental Info Re 990325 Request for Approval of Repair Plan for Feedwater Nozzle to safe-end Weld. Clarification Provided on Listed Three Items,Per 990329 Telcon with NRC1999-04-0101 April 1999 Provides Supplemental Info Re 990325 Request for Approval of Repair Plan for Feedwater Nozzle to safe-end Weld. Clarification Provided on Listed Three Items,Per 990329 Telcon with NRC PY-CEI-NRR-2383, Withdraws Inservice Exam Relief Requests IR-036 & IR-041, Submitted in Licensee to Nrc.Relief Request IR-041 Is Being Clarified with Respect to Alternative Proposed1999-04-0101 April 1999 Withdraws Inservice Exam Relief Requests IR-036 & IR-041, Submitted in Licensee to Nrc.Relief Request IR-041 Is Being Clarified with Respect to Alternative Proposed L-99-052, Forwards Info Concerning Dls Decommissioning Planning for Bvps,Units 1 & 2 & Pnpp Unit 1.DL Is Continuing to Utilize External Sinking Funds to Provide Financial Assurance for Nuclear Decommissioning Funding1999-03-30030 March 1999 Forwards Info Concerning Dls Decommissioning Planning for Bvps,Units 1 & 2 & Pnpp Unit 1.DL Is Continuing to Utilize External Sinking Funds to Provide Financial Assurance for Nuclear Decommissioning Funding ML20205H6171999-03-29029 March 1999 Forwards Util Consolidated Financial Statements for Twelve Month Period Ending 981231 & Internal Cash Flow Projection, Including Actual 1998 Data & Projections for 1999 PY-CEI-NRR-2377, Submits Decommissioning Repts for Bvps,Units 1 & 2,Davis- Besse Nuclear Power Station,Unit 1 & Perry Nuclear Power Plant,Unit 1,per 10CFR50.75(f)(1)1999-03-29029 March 1999 Submits Decommissioning Repts for Bvps,Units 1 & 2,Davis- Besse Nuclear Power Station,Unit 1 & Perry Nuclear Power Plant,Unit 1,per 10CFR50.75(f)(1) PY-CEI-NRR-2380, Requests Approval of Attached Repair Plan for Feedwater Nozzle to safe-end Weld N4C-KB.Proposed Weld Overlay Design & Installation Plan Details Included in Attachment 1. Attachment 2 Provides Applicable Preliminary Weld Procedure1999-03-25025 March 1999 Requests Approval of Attached Repair Plan for Feedwater Nozzle to safe-end Weld N4C-KB.Proposed Weld Overlay Design & Installation Plan Details Included in Attachment 1. Attachment 2 Provides Applicable Preliminary Weld Procedure 1999-09-09
[Table view] Category:UTILITY TO NRC
MONTHYEARPY-CEI-NRR-1226, Forwards LERs 90-018-00 & 90-019-00 for Plant1990-09-14014 September 1990 Forwards LERs 90-018-00 & 90-019-00 for Plant PY-CEI-NRR-1218, Responds to Generic Ltr 90-07 Re Proposed OL Exam Schedule, Operator Requalification Schedule & Generic Fundamental Exam Schedule for Plant.Licensee Will Continue to Train & Requalify Operators on Annual Basis1990-09-10010 September 1990 Responds to Generic Ltr 90-07 Re Proposed OL Exam Schedule, Operator Requalification Schedule & Generic Fundamental Exam Schedule for Plant.Licensee Will Continue to Train & Requalify Operators on Annual Basis ML20059E8741990-09-0505 September 1990 Forwards Fourteenth Quarterly Rept Cleveland Electric Illuminating Co Seismic Monitoring Network Jan-June 1990. Five Microevents Originating within Network & Three Other Events Originating Outside of Network Recorded PY-CEI-NRR-1216, Forwards Perry Nuclear Power Plant Unit 1 Semiannual Radioactive Effluent Release Rept,1990 Quarters 1 & 2. Response to Recommendations from NRC Re Range of Dilution Flows Will Be Submitted W/Next Effluent Rept1990-08-29029 August 1990 Forwards Perry Nuclear Power Plant Unit 1 Semiannual Radioactive Effluent Release Rept,1990 Quarters 1 & 2. Response to Recommendations from NRC Re Range of Dilution Flows Will Be Submitted W/Next Effluent Rept PY-CEI-NRR-1201, Forwards First Semiannual fitness-for-duty Rept for Plant for 900103-0630,per 10CFR26.71(d) Requirements1990-08-28028 August 1990 Forwards First Semiannual fitness-for-duty Rept for Plant for 900103-0630,per 10CFR26.71(d) Requirements PY-CEI-NRR-1211, Responds to NRC Re Violations Noted in Insp Rept 50-440/89-26.Corrective Actions Noted:Necessary Engineering Design Mods Which Evaluation Found Necessary Contained in DCP 90-00861990-08-23023 August 1990 Responds to NRC Re Violations Noted in Insp Rept 50-440/89-26.Corrective Actions Noted:Necessary Engineering Design Mods Which Evaluation Found Necessary Contained in DCP 90-0086 PY-CEI-NRR-1210, Proposes Date of 901207 to Provide Updated Status & Schedules,Per 900604 Request for Addl Info on Disposition of Bulletin 88-005 Product Forms Other than Pipe Fittings & Flanges Received at Plant1990-08-17017 August 1990 Proposes Date of 901207 to Provide Updated Status & Schedules,Per 900604 Request for Addl Info on Disposition of Bulletin 88-005 Product Forms Other than Pipe Fittings & Flanges Received at Plant PY-CEI-NRR-1208, Responds to NRC 900613 Request for Addl Info Re Inservice Insp Relief Requests.Util Revised Relief Requests PT-001 & PT-003 to Provide More Detailed Info & Has Withdrawn Relief Request IR-017.Revised Relief Requests & Summary Encl1990-08-10010 August 1990 Responds to NRC 900613 Request for Addl Info Re Inservice Insp Relief Requests.Util Revised Relief Requests PT-001 & PT-003 to Provide More Detailed Info & Has Withdrawn Relief Request IR-017.Revised Relief Requests & Summary Encl PY-CEI-NRR-1203, Provides Estimate Duration Schedule Re Installation of post-LOCA Neutron Monitoring Sys.Util Will Evaluate Effect of Resolution on Schedule & Confirm or Revise Installation Date as Needed,When Appeal & Criteria Questions Resolved1990-07-27027 July 1990 Provides Estimate Duration Schedule Re Installation of post-LOCA Neutron Monitoring Sys.Util Will Evaluate Effect of Resolution on Schedule & Confirm or Revise Installation Date as Needed,When Appeal & Criteria Questions Resolved PY-CEI-NRR-1200, Forwards Decommissioning Repts on Financial Assurance for Plant1990-07-25025 July 1990 Forwards Decommissioning Repts on Financial Assurance for Plant PY-CEI-NRR-1196, Forwards Rev 0 to Inservice Exam Program. Program Serves as 10-yr Rev to Inservice Insp Program Plan1990-07-24024 July 1990 Forwards Rev 0 to Inservice Exam Program. Program Serves as 10-yr Rev to Inservice Insp Program Plan PY-CEI-NRR-1205, Forwards LERs 90-014-00 & 90-015-00 for Plant1990-07-20020 July 1990 Forwards LERs 90-014-00 & 90-015-00 for Plant PY-CEI-NRR-1193, Responds to Generic Ltr 90-04 Re Implementation Status of Generic Safety Issues1990-06-29029 June 1990 Responds to Generic Ltr 90-04 Re Implementation Status of Generic Safety Issues ML20043H7371990-06-19019 June 1990 Forwards Individual Plant Exam Project Milestones.Milestones Subj to Change from Project or External Causes & Will Not Be Docketed Again Unless Significant Impact on Final Rept Contents or Submittal Date Exists ML20043H1711990-06-14014 June 1990 Discusses Util Radiological Safety Board of Ohio 900614 Comments on Notice of Opportunity for Hearing Re Rev to Previously Submitted License Amend Application.Licensees Would Welcome Opportunity to Consult W/State of Oh ML20043E5771990-06-0808 June 1990 Provides Info Re Implementation of Administrative Controls & Training Related to Use of Tech Spec 3.0.4.Util Intends to Implement Amend 30 Changes by 900618.All Licensed Operators Would Receive Training to Generic Ltr 87-09 ML20043B6981990-05-21021 May 1990 Forwards Info Re Activation Energy Values Utilized for Brand-Rex Co Cable at Facility.Encl Includes Rept, Qualification Tests of Electric Cables for Class 1E Svc in Nuclear Power Plants. ML20043A7041990-05-18018 May 1990 Forwards LERs 90-007-00 & 90-008-00 ML20043A4771990-05-16016 May 1990 Provides Justification Re Events Leading to Declaration of Alert on 900403 Due to Loss of Emergency Svc Water When Manway Flange Gasket Leak Developed at Pump a Discharge Strainer,In Response to Ocre 900406 Petition ML20042F4931990-05-0404 May 1990 Forwards Response to Generic Ltr 89-19 Re Reactor Vessel Overfill Protection ML20042F1961990-04-27027 April 1990 Informs of Adoption of Reorganization Plan Re Plants on 900424.Reorganization Will Make No Changes in Technical or Financial Qualifications for Plants.Application for Amends to Licenses Adding Company as Licensee Will Be Submitted ML20042F8051990-04-19019 April 1990 Forwards 1989 Financial Repts for Centerior Energy Corp, Cleveland Electric Illuminating Co & Toledo Edison Co & Supplemental Sec Form 10-K ML20012F3521990-04-0202 April 1990 Suppls Response to NRC 891215 Ltr Re Deviations Noted in Insp Rept 50-440/89-26.Corrective Actions:Profile of Heat Generation from Stored Fuel Vs Time for First Four Fuel Load Discharges Calculated & Design Changes Implemented ML20012E7811990-03-30030 March 1990 Confirms Basis for 890417 Submittal Re Station Blackout ML20012E9431990-03-26026 March 1990 Forwards Correction Pages to Semiannual Radioactive Effluent Release Rept 1989:Quarters 3 & 4. Changes Affect Liquid Effluent Data Only.Corrections Made Due to High Contribution of Sr-89 to Liquid Effluent Dose ML20012B9851990-03-0909 March 1990 Responds to Generic Ltr 90-01 Re Regulatory Impact Survey ML20012B5931990-03-0808 March 1990 Responds to Open Item Noted in Insp Rept 50-440/89-29. Corrective Actions:Plant Health Physicist & Operations Superintendent Met W/Health Physics & Operations Crews to Discuss Better Methods of Access Control ML20012A1561990-03-0202 March 1990 Forwards Semiannual Radioactive Effluent Release Rept for Jul-Dec 1989 & Addl Info in Response to NRC 890605 Ltr Re Offsite Dose Calculation Manual ML20012B9801990-03-0202 March 1990 Responds to 890810 Request for Addl Info on Util 890306 Tech Spec Change Request Re MSIV Leakage Limits.Change Does Not Request Increase in MSIV Leakage Beyond Currently Assumed in Updated SAR Chapter 15 ML20006E5131990-02-12012 February 1990 Lists Suggestions on Subj Areas to Be Addressed During Proposed 1990 Licensee Info Conference,In Response to 900117 Ltr ML20006E3211990-02-0909 February 1990 Discusses Implementation of Facility Revised Inservice Testing Program,Per Generic Ltr 89-04.Improvements Listed, Including sign-off Matrix Deletion at End of Instruction & Movement of Comments Section ML20006D0971990-02-0202 February 1990 Provides Update to 880721 Ltr Re Performance of Seismic Margin Evaluation.Util Awaiting Issuance of Individual Plant Exams of External Events Generic Ltr Prior to Proceeding W/ Seismic Margins Study ML20011E1841990-02-0202 February 1990 Forwards LERs 90-001 & 90-002 ML20006D5731990-01-30030 January 1990 Forwards Response to NRC 900112 Request for Addl Info Re 891121 Request for Exemption from Filing Requirements of 10CFR55.45(b)(2)(iii).Qualification Plan for Certification of Existing Simulator Approved in Aug 1987 ML20006B5131990-01-26026 January 1990 Responds to NRC Bulletin 89-002 Re Stress Corrosion Cracking of Type 410 Stainless Preloaded Bolting in Swing Check Valves.No Bolts Subj to Bulletin in Use at Facility ML20006B5571990-01-26026 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Procedures in Place for Underwater Insp of Intake Structures in Lake as Well as on-shore Pump Houses Annually ML20006A0781990-01-16016 January 1990 Responds to NRC 891215 Ltr Re Deviations Noted in Insp Rept 50-440/89-26.Corrective Actions:Profile of Heat Generation from Stored Fuel Vs Time for First Four Fuel Discharges Calculated & Updated SAR Change Request Will Be Submitted ML20005F9331990-01-0909 January 1990 Provides Inadvertently Omitted Inequality Signs from Util 891213 Response to NRC Bulletin 88-004 Re RCIC Pump Min Flow Limitations from Operating Instructions ML20042D5051990-01-0303 January 1990 Informs That Licensee Implemented fitness-for-duty Program at Plant on 900103,per 10CFR26.More Stringent cut-off Levels Imposed by Util than Required by 10CFR26 ML20005E1391989-12-28028 December 1989 Forwards Response to Generic Ltr 89-10 Re safety-related motor-operated Valve Testing & Surveillance.Each motor- Operated Valve Will Be Stroke Tested to Verify Valve Operability at no-pressure & no-flow Conditions ML19332F7941989-12-13013 December 1989 Responds to NRC 890531 Request for Addl Info Re NRC Bulletin 88-004, Safety-Related Pump Loss. Info Covers pump-to-pump Interaction That Could Result in Deadheading & Adequacy of Min Flow Bypass Lines Against Pump Damage ML19332E6921989-12-0707 December 1989 Responds to Violations Noted in Insp Rept 50-440/89-12 on 890627 & 1023.Corrective Actions:Flange Leakage of Relief Valve Repaired During First Refueling Outage & Design Mods to Enhance Testability of Gasketed Flanges Being Developed ML19332F3851989-12-0606 December 1989 Forwards Twelfth Quarterly Rept Cleveland Electric Illuminating Seismic Monitoring Network,Jul-Sept 1989. Rept Describes Area in Vicinity of Two Waste Injection Wells Located 3 Miles South of Plant Site ML19332E7971989-12-0101 December 1989 Responds to Generic Ltr 89-21 Re Implementation Status of USIs ML19332E3611989-11-28028 November 1989 Submits Supplemental Info Re Tech Spec Change Request Proposing Addition of Requirement to Verify Trip Setpoint of Intermediate Range Monitor Control Rod Block Instrumentation During Performance of Weekly Channel Functional Testing ML19332D5771989-11-27027 November 1989 Submits Supplementary Info to 880520 Tech Spec Change Request,Changing Setpoints for Turbine First Stage Pressure to Reflect Values Actually Measured During Startup Test Program at 40% of Rated Reactor Power.Table Encl ML19332C7341989-11-22022 November 1989 Forwards Executed Amend 3 to Indemnity Agreement B-98 ML19332C4041989-11-22022 November 1989 Responds to NRC 891025 Ltr Re Violations Noted in Insp Rept 50-440/89-23.Corrective Actions:Work Stopped on Div III & Restoration Began to Return to Operation & All Actions Required by Tech Spec 3.0.3 Completed ML19332C8321989-11-21021 November 1989 Forwards Inservice Insp Summary Rept NP0059-0001, Results for Outage 1 (1989),First Period,First Interval, Vols I & Ii,Covering Insp Period 871118-890805 ML19332C4531989-11-21021 November 1989 Requests Exemption from Requirements of 10CFR45.(b)(2)(iii), to Allow for Submittal of NRC Form 474, Simulation Facility Certification, After 910326 Deadline 1990-09-05
[Table view] |
Text
s ,o iliE C L EV E L A N D E L E Cin!C iL L U!?ll! ATIN G C O M P A N Y P o. Box 5000 m CLEVELAND oHlo 44101 e TELEPHONE (216) 622-9800 m ILLUMIN ATING BLOG e 55 PUBLIC SoVARE Datwyn R. Davidson "U " ' " "'" '"
VICE PRESIDENT 5ySTEM E NGINEERiNG AND CONSTRUCTION
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February 3, 1982 s'/ 0 Mr. A. Schwencer, Chief q N Licensing Branch No. 2 E D f .- o ' 'i ~ lP Division of Licensing v1 g'{3 ;; - ~'
,Q U. S. IMclear Regulatory Commission /
Washington, D. C. 20555 g s'(j t,g t/
Perry nuclear Power Plant Docket Hos. 50-440; 50-441 Response to Request for Meeting - Instrumentation and Control Systems
Dear Mr. Schwencer:
This letter and its attachment is submitted to provide draft respenses to several of the concerns identified in your letter dated November l'7,1981. A meeting was held on January 13 and 14 to discuss these concerns ard identify action items. Remaining agenda items will be addressed in future correspondence as agreed upon with the Instrumentation and Control Systems Branch reviewers.
Very Truly Yours, l l
- V V1 Dalwyn R. Davidson Vice President System Engineering and Construction DRD: clb cc: Jay Silberg, Esq.
M. Dean Houston Max Gildner, HRC Resident Inspector J. Mauck D
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6 421.02 Page 1 of 4 421.02 Several previously reviewed BWR installations, e.g., Grand Gulf, included a start-up transient monitoring system to-provide recordings of. selected parameters during the start-up and warranty testing. There is no information in the FSAR which describes this system. If this system, or any similar 4 system, is intended for use in the Perry units, provide the 7
following information:
- a. Identify all safety-related parameters which will be monitored with the transient monitoring system during start-up.
- b. For each safety parameter identified above, provide a concise description of how the associated circuitry merges or connects (either directly, or indirectly by means of isolation devices) with the circuitry associated with the transient monitoring system. Where appropriate, supplement this description with detailed electrical schematics.
- c. Describe provisions of the design to prevent-failures of this system from degrading safety-related systems.
Response
- 1. Perry intends to utilize the Emergency Response Information System C (ERIS) data acquisition system to monitor transients during startup. The ERIS is an integrated system that gathers the required plant data, stores and processes that data, generates visual displays for the operator and other personnel who need plant status information, provides printed records of transient events and has the capability of transmitting information to Technical Support Center and Emergency Operating Facility as described in NUREG 0696 " Functional Criteria for Emergency Response Facilities".
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- 2. The equipment monitoring safety related functions will be permanently installed to the same standards as all other plant equipment as described in .Section.8.3.1.4.l'and will be consistent with the separation
- requirements of Regulatory Guide 1.75.
- 3. The following parameters in safety-related systems will be monitored for startup transient testing.
System Parameters Neutron Monitoring System (NMS) APRM Output APRM Heat Flux APRM Flow Biased Rod Block LPRM Output Recire Sys Drive Flow Reactor Core Isolation Cooling RCIC Steamline AP (RCIC) RCIC Control Valve Position RCIC Stop Valve Position RCIC Steam Admission Valve Position RCIC Initiation RCIC Vessel ~ Injection Valve Position RCIC Discharge Flow RCIC Turbine Speed RCIC Turbine Controller Outputs RCIC Flow Controller Outputs RCIC Steam Pressure RCIC Discharge Pressure -
RCIC Suction Pressure l
RCIC Turbine Exhaust Pressure Low Pressure Core Spray (LPCS) LPCS Flow f
Residual Heat Removal (RHR) RHR Initiation i-RHR Heat -Exchanger Level i
, , - _ , , . . . ~ . - _ - _ , _ . . . ~ _- _. , _ . _ - . _ . . _ . _ - _ - , . _ - .. _ ._ . _ . _ . -
421.02-Page 3'of 4 System Parameters RHR Heat Exchanger Inlet Pressure RHR Heat Exchanger Pressure Controller
. Output RHR System Flow Rod Control ~ and Information Rod Scram Times -
(RCIS) Selected Rod Pilot Solenoid Nuclear Boiler / Nuclear Steam Outboard MSIV~ Position Supply Shutoff Inboard MSIV Position MSIV Isolation Initiation Signal Vessel Wide Range Level Standby Diesel Generator (D/G) D/G Initiation 4.16 kV-Power Distribution Emergency Bus'1 Status Emergency Bus 2 Status n
p ,[ Emergency Bus 3 Status x #
- g. Reactor Systems Reactor Vessel Pressure l Safety Relief Valve Initiation Containment. Isolation Logic Status L' Recirc. Pump Trip Breaker Status i-
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RPS Logic Status Jh , s "
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- 4. Isolation will be accomplished by converting signals for transmission via optical fiber cable. The optical isolation will be is . accomplished downstream of signal conditioning, multiplexing, and
.s analog-to-digital conversion. These remote multiplexers shall be
- ' - classified as divisional devices. Thus, within a given multiplexer only signals of one safety division will be connected. The signal conditioning and multiplexer unit will be qualified in accordance
., . with Regulatory Guides 1.89 and 1.100. The associated portion of the optical isolation shall be qualified in accordance with Regulatory Guides 1.75, 1.89, and 1.100.
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- 5. To maintain the signal conditioning and multiplexing equipment as divisional devices,'the power for these devices will be supplied from divisional power sources. In addition,_each signal input to the multiplexers will be individually conditioned and buffered from all other signals in the same multiplexer.
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. -421.06 Discuss conformance with the following TMI action items as required by NUREG-0737:
a .- II.D.3 Relief and safety valve position indication
- b. _II.F.2 - Inadequate Core Cooling
- c. II.K.3.18 ADS actuation
- d. II.K.3.21 - Restart of LPCS and LCPI
- e. II.K.3.22 - RCIC automatic-switchover
Response
- a. II.D.3 Relief and Safety Valve-Position Indication The SRV open/close monitoring system is'a single channel safety grade system consisting of a sensing element and a pressure switch connected to the discharge pipe at the downstream side of_the SRV
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discharge pipe. The electrical outputfof the pressure switch operates a relay which provioes input to the annunciator, process computer,._
and indicator lights. This system will be environmentally and seismically qualified. This system is identical to that recently-proposed by Grand Gulf and approved by NRC.
- b. This item will be addressed through LRG-II.
- c. This item will be addressed through LRG-II.
- d. II.K.3.21 Restart of LPCS and LPCI CEI endorses the BWR Owners Group position that providing controls to restart the low pressure core cooling systems would not enhance the overall BWR safety. However, CEI will modify the current HPCS control to provide automatic reset capability. This automatic reset modification of the HPCS resets the auto-initiation signal for low water level and blocks-the_ continuing auto-initiation signal for high drywell pressure.
This allows auto-restart of HPCS pump on low water level after the operator stopped the HPCS pcmp. The auto restart on high drywell
421.06 (Page 2 of 2) Cont'd pressure is blocked unless the high drywell pressure decreases below setpoint and again increases. Decrease in drywell pressure below trip level returns HPCS logic to original status,
- e. II.K.3.22 RCIC Automatic Switchover The RCIC pump suction automatic.switchover fion condensate storage tank to suppression pool has been committed by CEI for Perry and can be found in Perry FSAR Section 7.4.1.1.
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l 421.07 Describe the separation criteria for protection channel circuits, protection logic circuits, and non-safety related circuits. For example are channel circuits and logic circuits separated from one another?
Response
References to various FSAR sections of 8.3.1.4 were discussed in detail. The staff required no further action.
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Q 421.08 Table 7.1-2 of the FSAR states that the design of the Perry Reactor Protection System is similar to the design of the Grand Gulf Reactor Protection System. Provide a comparative discussion identifying specific differences between the two designs.
Response
Perry RPS design is functionally similar to Grand Gulf at this time. Some differences exist on turbine stop valve position sensors. The staff required no further action on this item.
. ~. 7 421.11 Revise the discussion concerning compliance with IEEE Standard 279 to verify that all portions of the RPS comply.
Response
The response to this question is provided in revised Section 7.2.2.2 attached.
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. ~. 7 The RPS is highly reliable and will provide a reactor scram in the event of anticipated operational occurrences.
7.2.2.2 Conformance to IEEE Standards The following is a discussion of conformance to those IEEE standards which apply specifically to the RPS system. Refer to Section 7.1.2.3 for a discussion of IEEE standards which apply equally to all safety related systems. The non-essential RPS power and its electrical protection assembly (EPA) are discussed in Section 8.3.1.1.5.1.
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- a. IEEE Standard 279 Criteria for Protection Systems for Nuclear Power )b Generating Stations - The RPS design complies with the requirements of IEEE-279. The following is a discussion of specific conformance.
- 1. General Functional Requirement (IEEE Standard 279, Paragraph 4.1)
The RPS automatically initiates the appropriate protective actions, whenever the conditions described in Section 7.2.1.1.b reach predetermined limits, with precision and reliability assuming the full range of conditions and performance discussed in Section 7.2.1.2.
- 2. Single Failure Criterion (IEEE Standard 279, Paragraph 4.2)
Each of the conditions (variables) described in Section 7.2.1.1.b is monitored by redundant sensors supplying input signals to redundant trip logics. Independence of redundant RPS equipment, cables, instrument tubing, etc. is maintained and single failure criteria preserved through the application of the PNPP separation criteria as described in Section 8.3.1 to assure that co single credible event can prevent the RPS from accomplishing its safety function.
- 3. Quality of Components and Modules (IEEE Standard 279, Paragraph 4.3)
For a discussion of the quality of RPS components and modules, refer to Section 3.11.
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421.12 In the discussion in Section 7.2.2.2 concerning conformance to Criterion 4.15 of IEEE Standard 279, the statement is made that there are no multiple setpoints within the RPS. Discuss the effects on RPS setpoints of mode switch operation.
Response
The response to this question is provided in revised Section 7.2.2.2.a.15 attached.
- 15. -Multiple' Set Points (IEEE Standard 279, Paragraph 4.15)
The reactor mode switch implements more restrictive scram trip setpoints when it is shifted from RUN to STARTUP. As the mode switch is moved to STARTUP . . .
(a) The APRM upscale neutron scram trip is replaced by the restrictive APRM setdown scram trip at 15 percent power.
-(b) The IRM range switch dependent scram trips are enabled.
Each IRM range switch enables successively more restrictive scram trip setpoints as it is ranged down.
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In addition to the mode switch dependent multiple setpoints, the flow 4
, channels which supply control and reference signals for the APRM 'M upscale thermal scram continually vary the scram setpoint as_ flow-changes. A sensed reduction in flow results in more restrictive scram trip setpoints.
The devices used to prevent improper use of the less restrictive setpoints (the mode switch, IRM range switches, the IRM and APRM signal conditioning equipment, and the flow channels) are designed in accordance with criteria regarding the performance and reliability of protection system e.quipment.
- 16. Completion of Protective Action Once it is Initiated (IEEE Standard 279; Paragraph 4.16) r Once the RPS trip logic has_been deenergized as a result of a trip channel becoming tripped, or the actuation of a manual scram switch, the trip-logic seal-ir contact opens _and completion of protection action is achieved without regard to the state of the initiating sensor trip, channel.
7.2-25
. . 7 After initial conditions (variable trip and logic deenergization) return to normal, deliberate operator action is required to return (reset) the RPS logic to normal (energ" zed).
- 17. Manual Initiation (IEEE Standard 279, Paragraph 4.17)
Refer to the discussion of Regulatory Guide 1.;22 in Section 7.2.2.3.a.
- 18. -Access to Set Point Adjustments, Calibration, and Test Points (IEEE Standard 279, Paragraph 4.18)
During reactor operation, access to set point or-calibration controls is not possible for the following RPS trip variables:
(a) Main steam line isolation valve' closure trip (b) Turbine stop valve closure trip (c) Turbine control valve fast closure trip Access to set point adjustments, calibration controls, and test points for all other RPS trip variables are under the administrative control of the' control room operator.
7.2-25a
. . R 421.13 Discuss the logic used for bypassing the turbine stop valve closure. Can a single failure in this pressure transmitter system cause a bypass of this closure to occur. Provide a similar discussion for the turbine control valves.
Response
The details of the bypass circuit werc discussed and the staff had no further questions.
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- 421.17 Identify any sensors or circuits used to provide input signals to the protection system which are located or routed through nonseismically qualified structures. This should' include sensors or circuits providing input for reactor trip, emergency safeguards equipment such as the Emergency Core Cooling system, and safety grade interlocks. - Verification should be provided that the sensors and circuits meet IEEE 279 and are seismically and environmentally qualified. Testing or analyses performed to insure that failures of non-seismic structures, mountings, etc., will not cause failures which could interfere with the operation of any other portion of the protection syrrem should be discussed.
Response
! Safety Related Inputs Routed in Non Seismic Structures I
RPS and ESF inputs routed in non seismic structures involve only inputs from
- sensors located on'the turbine or in the turbine building. These inputs go to the Reactor Protection System _(RPS) and the Main Steam Line Isolation Valve System (MSIV).
a.) RPS, trip reactor on turbine stop valve closure or control valve fast closure with bypass of both when turbine first stage pressure is below the equivalent of 40% power. Reference'Section_7.2.1.1; page 7.2-15 and 7.2-27.
b) 'MSIV, valve closure actuation on turbine inlet low pressure or main condenser low vacuum or high temperature in the steam line area of the turbine building. Reference Section 7.3.1.1.2, pages 7.3-16, 7.3-17 and 7.3-21.
All of these inputs and their circuits are treated in the same way all other safety related inputs and circuits are in terms of identification, location, mounting, and separation.
421.17 (Page 2 of 2) Cont'd In addition, all of these circuits, except the thermocouple circuits in the MSIV logic, are split with one conductor running on one side of the turbine building and the other running on the opposite side of the turbine building in order to minimize the possibility of a short due to damage resulting from a seismic event. The thermocouple circuits, because of the nature of their special wire requirement, are not split in their routing. However, since thermocouple wire carries no significant voltage or current and a thermocouple junction requires firm and substantial contact, the probability of cable damage resulting in a false thermocouple junction is small; the thermocouple temperature switch will trip upon detection of thermocouple open or burnout.
These, as well as all inputs to the RPS or MSIV logic, are isolated from other logic channels by trip units or relays, such that any failure of one channel will not prevent another channel from performing its function.
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q 421.18 Identify the physical location of-the equipment that actuates the reactor trip on turbine trip and indicate whether this equipment and associated circuitry meets the criteria applicable to equipment performing a safety function.
Response
The equipment that actuates reactor trip on turbine trip is located as shown:
Location Installation C71-N005A-D Turb. CV Fast Closure On Turb.40-021 (137D2407)(1)
C71-N006A-H Turb. Stop VLV Closure On Turb.40-853 (142D8582)(1)
C71-N052A-D Turb. 1st Stage Pressure 811-006,105 814-026, 027 (1) GE Turbine Drawing Numbers This equipment meets the criteria applicable to equipment performing safety functions that are described in Section 7.2.1.2, page 7.2-15; and-Section 7.2.2, pages 7.2-19, 7.2-22, 7.2-23, 7.2-25, and 7.2-28.
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m 421.20 'It has been noted during past reviews that' pressure switches or other devices were incorporated into the final actuation control. circuitry for large horsepower safety-related motors which are used to drive pumps. These switches or devices preclude automatic (safety signal) and manual operation of-the motor / pump combination unless permissive conditions, such
-as lube oil pressure, are. satisfied. Accordingly,; identify any safety-related motor / pump combinations which are used in the Perry design that operate as noted above. Also, describe the redundancy and diversity which are provided for the pressure switches or other permissive devices that are -
used in this manner.
Response
The response to this question has been provided as the response to Question 430.86 (submitted November 11, 1981).
7 421.22 The FSAR states that each ADS trip system has a time delay that can be reset manually to delay system initiation. Discuss the conditions under which the operator would reset the ADS timers.
Also, discuss the consequences of resetting the timers if the HPCS fails to start.
Response
The operation of the ADS trip system was discussed and the staff had no further questions.
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7 421.24 Discuss the' testing procedures used to demonstrate that the main. steam isolation valve closure time is within the 3 to 5' seconds assumed in Section 15.2.4.3.2.
Response
~ The cold stroke closure time of the main steam isolation valves will be set during preoperational testing.
The hot stroke closure time of-the main steam isolation valves will be determined during startup testing, as described in Section 14.2.12.2.22.1 of the FSAR. The Level 1 acceptance criteria are given in Section 14.2.12.2.22.ld.
Periodic testing of main steam isolation valves to determine closure times will be conducted. The acceptance criteria are that the main steam ,
isolation valves closure time, exclusive of electrical delay, shall be no faster than 3.0 seconds (average of the fastest valve in each steam line) -
and no slower than 5.0 seconds (each valve, not averaged).
421.25 Discuss how the Main Steamline Isolation Valve-Leakage Control System conforms to the requirements. of Paragraph 4.1 of IEEE..
Standard 279 concerning automatic initiation capability.
Response
This item was discussed, and it was concluded that auto-initiation of MSIV-LCS was not required. This system is designed to limit the long-term leakage through the MSIV af ter a LOCA is detected and, therefore, it is not needed to be manually initiated in the first 20 minutes after the MSIV auto closure.
This system is similar to Grand Gulf's.
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4 421.26 Discuss how the Suppression Pool Cooling Mode of the Residual Heat Removal System conforms to the requirements of Paragraph 4.1 of IEEE Standard 279 concerning automa;ic initiation capability.
Response
Operation of the system was discussed and the staff had no further questions.
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'421.30 The P&I diagrams for the Annulus Exhaust Gas Treatment-System are shown in Figure 6.5-1, Sheets.I and 2. However, Sheets 1 and 2 appear identical except for the FDIB and the
- FDRB valves. - Explain the significance of the two drawings.
Response
A discussion of instrument' tag numbers distinguishing units was held. 'The staff required no further action on this item.
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7 421.34 Table 7.1-2 of the FSAR identifies many ESF systems that are-similar to the design of the Grand Gulf ESF systems. Provide a comparative discussion identifying specific differences between designs of similar systems.
Response-The ESF Systems in the NSSS scope are functionally similar to Grand Gulf at this time.
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421.50 Using detailed system schematics, describe'the implementation of the bypassed and inoperable status indication provided for engineered safeguards features. Discuss how the design of the bypass.and inoperable status indication systems comply with positions B1 through B6 of Branch Technical Position ICSB No. 21.
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Response
This item was discussed and the staff required no further action.
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m 421.52 Why are the indicator lights or annunciators required by R. G. 1.47 not included in Table 7.5-1 or any other table in Section 7.57
Response
This item was discussed and the staff required no further action.
t 421.57- In the discussion concerning the leak detection instrumentation for fission product monitoring, Section 5.2.5.2.1, reference is made to Section 7.6. However, no information is provided in Section 7.6 relative to fission product monitoring. Discuss this instrumentation and the need to include a description of-it'in Section 7.6.
Response
The' leak detection instrumentation for fission product monitorin?, is discussed in Section 12.3.4. The correct reference is provided in revised Section 5.2.5.2.1, attached. .
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- d. Temperature Measurement.
The ambient temperature within the drywell is monitored by four single element thermocouples located equally spaced in the vertical direction within the drywell. An abnormal increase in drywell temperature could indicate a leak within the drywell. In addition, the drywell exit end of the containment penetration guard pipe for the main steam line is also monitored for abnormal temperature rise caused by leakage from the main steam line. Ambient temperatures within the drywell are recorded and alarmed on the LD&IS (Leakage Detection and Isolation System) control room panel.
- e. Fission Product Monitoring.
1 The drywell air sampling system is used along with the temperature, pressure, and flow variation method described above to detect leaks in the nuclear system process barrier. The system continuously monitors the drywell and drywell atmosphere for airborne radioactivity (iodine, noble gases and particulates). The sample is drawn from the ventilation.
exhaust of the containment and drywell. A sudden increase of activity, h
which may be attributed to steam or reactor water leakage, is annunciated %
v in the control room. (C2e Section 12.3.4). 4
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- f. Drywell Pressure Measurement.
The drywell pressure varies slightly during reactor operation and is monitored by pressure sensors. The' pressure fluctuates slightly as result of barometric pressure changes and outleakage. A pressure rise
, above the normally indicated values will indicate a possible leak within the drywell. Pressure exceeding the preset. values will be annunciated in the main control room and safety action will be automatically initiated, i
. g. Reactor Vessel Head Seal.
The reactor vessel head closure is provided with double seals with a leak of f connection between seals that is piped through a normally closed 5.2-48 4
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~1 421.62 Discuss the safety aspects of the Perry design for the following trips and interlocks:
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- a. Recirculation flow ntrol valve motion interlocks;-
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- b. Low reactor vessel-level and high vessel pressure recirculation pump trips;
- c. High reactor vessel water level trips for the feedwate'r pumps and plant turbine.
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Response
- a. Recirculation Flow Control Valve Motion Interlocks i
These interlocks are not safety grade. The. failures of the recirculation ,
flow control are analyzed in Chapter 15.
- b. Recirculation Pump Trip .
The trip on Low Vessel Water level and high vessel pressure is.the ATUS trip utilizing redundant sensing logic developed to trip a single breaker and is not safety grade. This is consistent with the HatchJVPWS fix.
- c. The high vessel water level trips for the feedwater pumps and plant turbine are not safety grade and are intended only to protect the turbine.
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c 421.63 :/ , ' Identify the non-safety grade equipment .used to mitigate the 7
/ effects of Anticipated Transients Without' Scram'(ATWS).
Inciude ' a discussion of the ATWS . recirculation pump trip -
. n- '(ATVSRPT).
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- Response
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G The enrrent equipment used in Perry to mitigate the effects of ATWS is .imilar i to the so-called Hatch fix 'which consists of non-safety grade circuitry- that
~ ' trips the recirculation pumps without transfer to the low frequency motor
' generator set' upon receipt of a high reactor pressure or low level (2) vessel 1
water level signals. This fix is similar to what Grand Gulf has. at this time.
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